The invention disclosed herein is generally related to radiation measuring instruments and methods. More particularly, this invention is related to apparatus for measuring the radiation emitted from spent nuclear fuel assemblies for the purpose of determining the burnup, or extent of consumption, and the related fissile content of the nuclear fuel contained therein.
The problem to which the present invention is directed is the in situ inspection of spent nuclear reactor fuel that is stored underwater in fuel storage ponds. Such inspection is an important element of the international effort to prevent wrongful diversion of special nuclear material that could be utilized to make nuclear weapons. There has not previously been available a simple yet efficient device for conducting quick on-site inspections to verify the operator-declared burnup and relate fissile content of spent fuel in storage.
Nuclear reactor fuel consists essentially of uranium which has been enriched in the fissile uranium isotope U.sup.235. The fuel is ordinarily contained in what are referred to as nuclear fuel assemblies. Such fuel assemblies are the basic fuel units that are loaded into a reactor core and subsequently removed at a later date when the fissile material has been consumed to a level at which it is no longer practical to extract additional energy. Each fuel assembly consists of a bundle of spaced-apart, parallel fuel rods, which are metal tubes loaded with the nuclear fuel.
As the nuclear fuel is "burned" in the reactor, the fissile U.sup.235 is progressively consumed by the induced fission process. However, not all of the fissile U.sup.235 is consumed during the period the fuel assembly is in the reactor. Induced fission effectively ceases when the assembly is removed from the reactor, so that there remains some unburned U.sup.235 in the spent fuel that is removed from the reactor. Additionally, the fission process produces other fissile materials, primarily plutonium-239, which accumulate in the fuel and remain there once the fuel assembly is removed from the reactor. Both U.sup.235 and Pu.sup.239 are special nuclear materials that must be safeguarded and accounted for after the spent fuel is removed from the reactor.
Since the induced fission process effectively stops when the spent fuel is removed from the reactor, the concentrations of U.sup.235 and Pu.sup.239 remain substantially constant thereafter. These isotopes gradually decay by spontaneous fission and alpha decay, but the rate of this decay is so slow as to be negligible for all practical purposes.
The extent of consumption of the U.sup.235 originally present in the fuel is called the "burnup" of the fuel. The actual burnup of the spent fuel taken from a given fuel assembly is a function primarily of the power level at which the reactor was operated, integrated over the period of time the assembly was in the reactor. Accordingly, the burnup of the irradiated fuel, also called the exposure of the fuel, is ordinarily given in units of megawatt-days per metric ton of uranium metal, or MWd/tU. Burnup is also frequently given in units of gigawatt-days per metric ton, or GWd/tU. In a typical commercial power reactor a fuel assembly is subjected to an exposure on the order of 40,000 MWd/tU, or 40 GWd/tU.
In addition to the fissile materials in the irradiated fuel, there are large amounts of radioactive nonfissile fission products, which are byproducts of the fission process. These fission products include a host of radioactive nuclides which decay at various decay rates. Some of the fission products produce high levels of radiation, and it is for this reason that the spent fuel assemblies must be stored in a shielded and isolated environment for years after they are removed from a reactor core. This is normally done by storing the spent fuel assemblies underwater in specially designed storage ponds.
The radioactive fission products can be chemically separated from the relatively smaller amounts of fissile materials by established reprocessing methods. The fissile materials separated and concentrated by such reprocessing are relatively less radioactive than the fission products, and can be safely handled, stored and safeguarded. However, reprocessing of commercial fuel has not been implemented on a large scale because of concern that small amounts of the purified fissile materials might be wrongfully diverted without detection during the reprocessing of large amounts of fuel. As a result, large amounts of spent nuclear fuel have accumulated in storage ponds around the world, and will continue to accumulate in the forseeable future. The presence of this accumulated fuel presents a danger, however, because reprocessing of limited amounts of spent fuel, by a sophisticated entity having access to the fuel, could theoretically be carried out on a relatively small scale for the purpose of obtaining sufficient fissile material to fabricate nuclear weapons.
Accordingly, international agreements have been entered into, under which it is the task of the International Atomic Energy Agency to periodically inspect spent fuel assemblies at storage sites around the world for the purpose of verifying that the stored fuel has not been altered or tampered with. It is intended that such inspection will deter any attempt to wrongfully divert the spent fuel or its fissile contents, or at least to provide prompt detection of any such diversion that may be accomplished. More specifically, it is the object of such inspection to directly verify the fissile material content of the spent fuel in storage.
Since there may be hundreds of fuel assemblies submerged in a fuel storage pond, and many such ponds to be inspected, it has been sought to provide a method and apparatus for conducting rapid yet accurate analyses of spent fuel assemblies while they are underwater.
Gamma ray analysis of irradiated fuel assemblies has been considered as a simple way to inspect such assemblies. This may be accomplished by immersing a conventional gamma ray ionization chamber, for example, in a fuel pond and taking a reading from a position adjacent to a fuel assembly. A closely related method which has been employed is to measure the Cerenkov light emitted from the water around a fuel assembly, which is generated by gamma radiation from the assembly. However, the latter method must sometimes be conducted in the dark, thereby requiring the inspector to operate in an awkward and possibly unsafe manner. More importantly, however, gamma measurements alone cannot unequivocally distinguish radioactive fission products containing fissile materials from fission products from which the fissile materials have been removed, or from radioactive activation products that may have been substituted for the irradiated fuel.
Passive neutron analysis of spent fuel assemblies is one approach to this problem, since it may be utilized in a nondestructive in situ method which provides a reliable determination of the fissile content of spent fuel assemblies. This approach is complicated, however, by the fact that some storage ponds contain dissolved boron salts to absorb neutrons emitted from the fuel. More specifically, storage ponds of pressurized water reactors ordinarily contain boron at varying concentrations, whereas storage ponds associated with boiling water reactors do not contain boron.